The following relates to the nuclear reactor arts, electrical power generation arts, nuclear safety arts, and related arts.
Nuclear reactors employ a reactor core comprising a mass of fissile material, such as a material containing uranium oxide (UO2) that is enriched in the fissile 235U isotope. Primary coolant water, such as light water (H2O) or heavy water (D2O) or some mixture thereof, flows through the reactor core to extract heat for use in heating secondary coolant water to generate steam or for some other useful purpose. For electrical power generation, the steam is used to drive a generator turbine. In thermal nuclear reactors, the primary coolant water also serves as a neutron moderator that thermalizes neutrons, which enhances reactivity of the fissile material. Various reactivity control mechanisms, such as mechanically operated control rods, chemical treatment of the primary coolant with a soluble neutron poison, or so forth are employed to regulate the reactivity and resultant heat generation. In a pressurized water reactor (PWR), the primary coolant water is maintained in a superheated state in a sealed pressure vessel that also contains the reactor core. In the PWR, both pressure and temperature of the primary coolant water are controlled.
To extract power from the PWR or other nuclear reactor, secondary coolant water is flowed in thermal communication with the primary coolant water. A steam generator device is suitably used for this thermal exchange. In the steam generator, heat (i.e., energy) is transferred from the reactor core to the secondary coolant water via the intermediary of the primary coolant water. This heat converts the secondary coolant water from liquid water to steam. The steam is typically flowed into a turbine or other power conversion apparatus that makes practical use of the steam power. Viewed another way, the steam generator also serves as a heat sink for the primary coolant.
The steam generator may, in general, be located external to the pressure vessel, or internal to the pressure vessel. A PWR with an internal steam generator is sometimes referred to as an integral PWR, an illustrative example of which is shown in Thome et al., “Integral Helical Coil Pressurized Water Nuclear Reactor”, U.S. Pub. No. 2010/0316181 A1 published Dec. 16, 2010 which is incorporated herein by reference in its entirety. This publication discloses a steam generator employing helical steam generator tubing; however, other coil geometries including straight (e.g., vertical) steam generator tubes are also known. This publication also discloses an integral PWR in which the control rod drive mechanism (CRDM) is also internal to the pressure vessel; however, external CRDM designs are also known. Some illustrative examples of internal CRDM designs include: Stambaugh et al., “Control Rod Drive Mechanism for Nuclear Reactor”, U.S. Pub. No. 2010/0316177 A1 published Dec. 16, 2010 which is incorporated herein by reference in its entirety; and Stambaugh et al., “Control Rod Drive Mechanism for Nuclear Reactor”, Intl Pub. WO 2010/144563 A1 published Dec. 16, 2010 which is incorporated herein by reference in its entirety.
During normal PWR operation, the primary coolant is subcooled and is at both elevated temperature and elevated pressure. For example, one contemplated integral PWR is designed to operate with the primary coolant at a temperature of greater than 300° C. and a pressure of about 2000 psia. These elevated conditions are maintained by emitted by the radioactive nuclear reactor core. In various abnormal event scenarios, this radioactivity can increase rapidly, potentially leading in turn to rapid and uncontrolled increase in primary coolant pressure and temperature. For example, in a “loss of heat sink event” the secondary coolant flow in the steam generator fails, leading to loss of heat sinking provided by the secondary coolant. In a scram failure, the control rod system is compromised such that the control rods may be unable to “scram”, that is, be released to fall into the reactor core, to provide rapid shutdown. While a scram failure may not cause immediate core heating, the loss of this safety system typically calls for immediate shutdown of the reactor. In a loss of coolant accident (LOCA), a rupture in the pressure vessel allows some of the primary coolant to be released under pressure from the pressure vessel. The released primary coolant generally expands as steam outside of the pressure vessel. A LOCA introduces numerous potential safety issues such as a possible release of radioactivity, emission of hot steam, and so forth. The loss of coolant as the reactor depressurizes can result in a condition where there is insufficient coolant left in the reactor vessel to cool the core. The resulting fuel damage releases fission products contained with the fuel.
In view of such concerns, a PWR typically has an external containment structure to contain any release of primary coolant in a LOCA. The PWR also typically has an associated emergency core cooling system (ECCS) that is designed to respond to an abnormal condition by bringing about rapid cooling of the reactor core, suppressing any concomitant pressure increase, and recapturing any released primary coolant steam. The ECCS should operate in a failsafe manner. However, designing the ECCS to provide failsafe operation for a range of potential abnormal conditions such as loss of heat sinking, scram failure, or LOCA is difficult.
Disclosed herein are improvements that provide various benefits that will become apparent to the skilled artisan upon reading the following.